Development of Thermal-Hydraulic Analysis Code for Plate Type Fuel Reactor

2010 ◽  
Author(s):  
Lei Li ◽  
Zhijian Zhang
Keyword(s):  
Hydraulic Analysis ◽  
Fuel Reactor ◽  
Thermal Hydraulic Analysis ◽  
Type Fuel ◽  
Plate Type

Development of Real-Time Thermal-Hydraulic Analysis Code for Plate Type Fuel Reactor

2010 ◽  
Author(s):  
Lei Li ◽  
Zhijian Zhang
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Real Time ◽  
Transient Flow ◽  
Rectangular Channel ◽  
Hydraulic Analysis ◽  
Fuel Reactor ◽  
Thermal Hydraulic Analysis ◽  
Type Fuel ◽  
Plate Type

A multi-channel model thermal-hydraulic analysis code in real-time for plate type fuel reactor is developed in this paper. In this code, every fuel assembly in reactor is divided into a subchannel. A series of reasonable mathematical and physical model are set up based on the structure and operational characteristics of plate type fuel core. As for the choice of flow friction and heat transfer models, all possible flow regimes which include the laminar flow, transient flow and turbulent flow, and heat transfer regimes which include single liquid phase heat transfer, sub-cooled boiling, saturation boiling, film boiling and single vapor phase heat transfer, are considered. The correlations and constitutive equations used in the code are fit for the rectangular channel. Look-up table method is used to calculate the properties of water and steam. The code has been loaded on the real-time simulation supporting system SimExec. The reactivity insertion accident and loss of flow accident, which has been defined in the IAEA 10MW MTR benchmark program, were calculated by the code in this paper for validation. Furthermore, the steady state of CARR (China Advanced Research Reactor) is analyzed by this code. The detailed flow distribution in each fuel assembly is obtained. The temperature of coolant, quality, void fraction, DNBR in each subchannel is calculated. The results show that the recently developed code can be used for real time thermal hydraulic analysis of plate type fuel reactor.

Thermal-Hydraulic Analysis Under LOFA Hypothesis of a Plate-Type Fuel Surrounded by Two Water Channels Using Relap5 Code

2015 ◽  
Author(s):  
Itamar Iliuk ◽  
José Manoel Balthazar ◽  
Angelo Marcelo Tusset ◽  
José Roberto Castilho Piqueira
Keyword(s):  
Water Channels ◽  
Hydraulic Analysis ◽  
Thermal Hydraulic Analysis ◽  
Type Fuel ◽  
Plate Type ◽  
Relap5 Code

scholarly journals Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

2018 ◽  
Vol 2018 ◽  
pp. 1-17
Author(s):  
Duvan A. Castellanos-Gonzalez ◽  
João Manoel Losada Moreira ◽  
José Rubens Maiorino ◽  
Pedro Carajilescov
Keyword(s):  
Steady State ◽  
Pressure Drop ◽  
Fuel Assembly ◽  
Nuclear Reactors ◽  
Nucleate Boiling ◽  
Hydraulic Analysis ◽  
Fuel Assemblies ◽  
Thermal Hydraulic Analysis ◽  
Type Fuel ◽  
Plate Type

This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.

scholarly journals Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

2016 ◽  
Vol 8 (1) ◽  
pp. 168781401562636 ◽  
Author(s):  
Itamar Iliuk ◽  
José Manoel Balthazar ◽  
Ângelo Marcelo Tusset ◽  
José Roberto Castilho Piqueira
Keyword(s):  
Water Channels ◽  
Hydraulic Analysis ◽  
Partial Loss ◽  
Thermal Hydraulic Analysis ◽  
Type Fuel ◽  
Plate Type ◽  
Relap5 Code

scholarly journals Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core

2020 ◽  
Vol 2020 ◽  
pp. 1-12 ◽  
Author(s):  
Linrong Ye ◽  
Mingjun Wang ◽  
Xin’an Wang ◽  
Jian Deng ◽  
Yan Xiang ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Coupling Method ◽  
High Fidelity ◽  
Monte Carlo Code ◽  
Coupling Analysis ◽  
Fuel Reactor ◽  
Nuclear Reactor Core ◽  
Type Fuel ◽  
Plate Type

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

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